This invention generally relates to an apparatus and process for electrolytically removing radioactive ions from a decontamination solution in order to regenerate the same. The invention also reduces the ions to small volume of metals and ash which are easily encapsulated in a cementitious matrix without the formation of liquid radioactive wastes.
Various methods for removing the radioactive ions from chemical decontamination solutions are known in the prior art. However, before these removal methods are discussed, a brief description of the purpose and composition of the decontamination solutions themselves will be given so that the significance of the invention may be more easily appreciated.
Generally, the decontamination solutions that the invention pertains to are used to remove magnetite deposits that gradually build up in the water conduits which form the cooling system of nuclear reactors. The magnetite deposits contain radioactive metals, and the removal of these deposits is necessary to safely maintain and repair such cooling systems. These deposits are typically removed by first treating them with an oxidizing solution, such as one containing an alkaline permanganate, to remove the chromium therefrom. This step renders the magnetite much more dissolvable in an acidic solution. The chromium-depleted magnetite deposits are then treated with a decontamination solution, which is an aqueous solution of a chelate, such as ethylenediaminetetraacetic acid (EDTA), and a solubilizing agent, such as a mixture of oxalic acid and citric acid. Other chelates which may be used include oxybis (ethylenedraminetetracetic acid) (EEDTA), and nitrilotriacetic acid (NTA). The chelate forms a complex with the radioactive metal ions from the magnetite deposits and solubilizes them, thus preventing them from precipitating out of the solution at another location in the cooling system.
Ultimately, the radioactive metal ions captured by the chelate must be removed from the decontamination solution in order to regenerate the solution. Moreover, the removed radioactive ions must then be put into a form which is easily and inexpensively disposable. One prior art method for removing the ions from the decontamination solution involved circulating the solution between the cooling system of the nuclear reactor and a cation exchange resin. The chelated metal ions were deposited on the cation exchange resin, freeing the chelates to solubilize additional metal ions in the deposit. However, since both the chelates and the cation exchange resin compete for the metal ions, the ions do not readily leave the chelate and attach themselves to the ion exchange column. As a result, long resin contact times are required, and the resulting column effluent may include relatively large amounts of liquid wastes containing high concentrations of radioactive ions. Hence, in addition to taking a lengthy amount of time to effect decontamination, this ion exchange process creates a radioactive liquid effluent that is relatively difficult and expensive to dispose of.
To solve these problems, the inventors developed an electrolytic method for removing these metal ions from the chelates in the decontamination solutions. This new method is described in and claimed U.S. Pat. No. 4,537,666 issued Aug. 27, 1985, and assigned to the Westinghouse Electric Corporation. Generally speaking, this process passes the decontamination solution through an electrode formed by a stainless steel or copper mesh in order to plate the ions out. When the electrode becomes completely plated out and hence spent, it is replaced with a fresh electrode.
However, while the process described and claimed in this patent represents a substantial advance in the art, the applicants have observed that there is room for improvement on several of the aspects of this invention. For example, of the volume of solid waste produced by this process (i.e., the spent and plated electrode) more than 99% is non-radioactive metal. Since the cost of disposal is directly proportional to the volume of the radioactive waste, the fact that only a very tiny volume of the metal of on the spent electrodes is radioactive is an unfortunate inefficiency. A second undesirable characteristic of the prior art electrolytic process is the fact that of the metallic electrodes actually used, some were prone to corrosion (such as copper) while others (such as stainless steel) were found to have short lifespans due to passivation. Still another undesirable characteristic of the prior art electrolytic process was the fact that the electrodes used therein had no ability to filter or adsorb impurities (such as lubricating oils and other hydrophobic compounds) which are often present in at least trace amounts in the decontamination solutions. The ion exchange column used before in the prior art did offer some filtration and adsorption capability in this regard, and while the more recently developed electrolytic process is, on the balance, far superior to the ion exchange method, the loss of this filtration and adsorption capability represents the loss of a significant advantage.
Clearly, there is a need for an improved process and apparatus for removing the metal ions from decontamination solutions which retains all of the advantages of both the prior art electrolytic and ion exchange processes, but which produces no liquid radioactive wastes. Ideally, such a process should utilize components having a long lifespan, and produce solid wastes of greatly reduced volume. Moreover, such a process should retain the filtration and adsorption advantages associated with the prior art ion exchange columns.